[Federal Register Volume 72, Number 19 (Tuesday, January 30, 2007)]
[Notices]
[Pages 4304-4312]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-1259]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission publish notice of any amendments
issued, or proposed to be issued and grants the Commission the
authority to issue and make immediately effective any amendment to an
operating license upon a determination by the Commission that such
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from January 5, 2007 to January 18, 2007. The
last biweekly notice was published on January 16, 2007 (72 FR 1779).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3)
[[Page 4305]]
involve a significant reduction in a margin of safety. The basis for
this proposed determination for each amendment request is shown below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rulemaking,
Directives and Editing Branch, Division of Administrative Services,
Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the Commission's
Public Document Room (PDR), located at One White Flint North, Public
File Area O1F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. The filing of requests for a hearing and petitions for leave
to intervene is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and
[[Page 4306]]
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of amendment request: November 16, 2006.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Section 3.6.5.1, ``Drywell,''
Surveillance Requirement (SR) 3.6.5.1.3 to delay the performance of the
next drywell bypass leakage rate test (DBLRT) from the current
requirement of ``November 23, 2008'' to ``prior to startup from the
C1R12 refueling outage'' which is currently scheduled for January 2010.
This request would also revise TS 5.5.13, ``Primary Containment Leakage
Rate Testing Program,'' to delay the performance of the next primary
containment Type A integrated leak rate test (ILTR) from the current
requirement of ``no later than November 23, 2008'' to ``prior to
startup from the C1R12 refueling outage.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change will revise TS 3.6.5.1, ``Drywell,'' SR
3.6.5.1 .3 to defer the performance of the next DBLRT to prior to
startup from the C1R12 refueling outage. This request will also
revise CPS TS 5.5.13, ``Primary Containment Leakage Rate Testing
Program,'' to reflect a one-time deferral of the primary containment
Type A test to prior to startup from the C1R12 refueling outage. The
current Type A test and DBLRT interval of 15 years, based on past
performance, would be extended on a onetime basis to 16.25 years
(i.e., approximately 15 years plus 15 months) from the last Type A
test and DBLRT.
The drywell houses the reactor pressure vessel, the reactor
coolant recirculation loops, and branch connections of the Reactor
Coolant System (RCS), which have isolation valves at the primary
containment boundary. The function of the drywell is to maintain a
pressure boundary that channels steam resulting from a Loss of
Coolant Accident (LOCA) to the suppression pool, where it is
condensed. Air forced from the drywell is released into the primary
containment through the suppression pool. The suppression pool is a
concentric open container of water with a stainless steel liner that
is located at the bottom of the primary containment. The suppression
pool is designed to absorb the decay heat and sensible heat released
during a reactor blowdown from safety/relief valve (SRV) discharges
or from a LOCA.
The function of the Mark III containment is to isolate and
contain fission products released from the RCS following a design
basis LOCA and to confine the postulated release of radioactive
material to within limits. The test interval associated with the
drywell bypass leakage and Type A testing is not a precursor of any
accident previously evaluated. Therefore, extending these test
intervals on a one-time basis from 15 years to 16.25 years does not
result in an increase in the probability of occurrence of an
accident. The successful performance history of the drywell bypass
leakage and Type A testing provides assurance that the CPS drywell
and primary containment will not exceed allowable leakage rate
values specified in the TS and will continue to perform its design
function following an accident. The risk assessment of the proposed
changes has concluded that there is an insignificant increase in
total population dose rate and an insignificant increase in the
conditional containment failure probability.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes for a one-time extension of the DBLRT and
Type A test will not affect the control parameters governing unit
operations or the response of plant equipment to transient and
accident conditions. The proposed changes do not introduce any new
equipment or modes of system operation. No installed equipment will
be operated in a new or different manner. As such, no new failure
mechanisms are introduced.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
CPS is a General Electric BWR/6 plant with a Mark III
containment system. The Mark III containment design is a single-
barrier pressure containment and a multi-barrier fission containment
system consisting of the drywell and primary containment. The
drywell houses the reactor pressure vessel, the reactor coolant
recirculation loops, and branch connections of the RCS, which have
isolation valves at the primary containment boundary. The function
of the drywell is to maintain a pressure boundary that channels
steam from a LOCA to the suppression pool, where it is condensed.
The suppression pool is an annular pool of demineralized water
between the drywell and the outer primary containment boundary. This
pool covers the horizontal vent openings in the drywell to maintain
a water seal between the drywell interior and the remainder of the
containment volume. The primary containment consists of a steel-
lined, reinforced concrete vessel, which surrounds the RCS and
provides an essentially leak-tight barrier against an uncontrolled
release of radioactive material to the environment. Additionally,
the containment structure provides shielding from the fission
products that may be present in the primary containment atmosphere
following accident conditions. The primary containment is penetrated
by access, piping and electrical penetrations.
The integrity of the drywell is periodically verified by
performance of the DBLRT. This test ensures that the measured
drywell bypass leakage is bounded by the safety analysis
assumptions. The drywell integrity is further verified by a number
of additional tests, including drywell airlock door seal leakage
tests, overall drywell airlock leakage tests and periodic visual
inspections of exposed accessible interior and exterior drywell
surfaces. Additional confidence that significant degradation in the
drywell leaktightness has not developed is provided by the periodic
qualitative assessment of drywell performance.
The integrity of the primary containment penetrations and
isolation valves is verified
[[Page 4307]]
through Type B and Type C local leak rate tests (LLRTs) and the
overall leak-tight integrity of the primary containment is verified
by a Type A integrated leak rate test (ILRT) as required by 10 CFR
50, Appendix J. These tests are performed to verify the essentially
leak-tight characteristics of the primary containment at the design
basis accident pressure. The proposed changes for a one-time
extension of the drywell bypass leakage and Type A tests do not
affect the method for drywell or containment testing or the test
acceptance criteria.
AmerGen has conducted a risk assessment to determine the impact
of a change to the CPS Type A ILRT and DBLRT schedule from the
originally licensed baseline frequency of three tests in 10 years to
one test in 15 years plus 15 months (i.e., approximately 16.25
years) for the risk measures of Large Early Release Frequency (i.e.,
LERF), Population Dose, and Conditional Containment Failure
Probability (i.e., CCFP). This assessment indicated that the
proposed CPS interval extension has a small change in risk to the
public and is an acceptable plant change from a risk perspective.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, AmerGen concludes that the proposed
amendment presents no significant hazards consideration under the
standards set forth in 10 CFR 50.92, ``Issuance of amendment,''
paragraph (c), and, accordingly, a finding of ``no significant
hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Assistant General
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett
Square, PA 19348.
NRC Branch Chief: Michael L. Marshall, Jr.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: November 2, 2006.
Description of amendment request: The proposed amendment would
modify requirements for inoperable snubbers consistent with the
Technical Specification Task Force 372, Revision 4.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated.
The proposed change allows a delay time for entering a supported
system technical specification (TS) when the inoperability is due
solely to an inoperable snubber if risk is assessed and managed. The
postulated seismic event requiring snubbers is a low-probability
occurrence and the overall TS system safety function would still be
available for the vast majority of anticipated challenges.
Therefore, the probability of an accident previously evaluated is
not significantly increased, if at all. The consequences of an
accident while relying on allowance provided by proposed LCO
[limiting condition for operation] 3.0.8 are no different than the
consequences of an accident while relying on the TS required actions
in effect without the allowance provided by proposed LCO 3.0.8.
Therefore, the consequences of an accident previously evaluated are
not significantly affected by this change. The addition of a
requirement to assess and manage the risk introduced by this change
will further minimize possible concerns. Therefore, this change does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility
of a New or Different Kind of Accident From Any Previously
Evaluated.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Allowing delay times for entering supported system TS when
inoperability is due solely to inoperable snubbers, if risk is
assessed and managed, will not introduce new failure modes or
effects and will not, in the absence of other unrelated failures,
lead to an accident whose consequences exceed the consequences of
accidents previously evaluated. The addition of a requirement to
assess and manage the risk introduced by this change will further
minimize possible concerns. Thus, this change does not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in [a] Margin of Safety.
The proposed change allows a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
snubber, if risk is assessed and managed. The postulated seismic
event requiring snubbers is a low-probability occurrence and the
overall TS system safety function would still be available for the
vast majority of anticipated challenges. The risk impact of the
proposed TS changes was assessed following the three-tiered approach
recommended in RG [Regulatory Guide] 1.177. A bounding risk
assessment was performed to justify the proposed TS changes. This
application of LCO 3.0.8 is predicated upon the licensee's
performance of a risk assessment and the management of plant risk.
The net change to the margin of safety is insignificant. Therefore,
this change does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: J. M. Fulton, Esquire, Assistant General
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth,
Massachusetts, 02360-5599
NRC Branch Chief: Richard Laufer.
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of amendment request: November 7, 2006.
Description of amendment request: The proposed change revises
Technical Specification (TS) Surveillance Requirement (SR) 3.4.3.1 to
increase the allowable as-found main steam safety valve (MSSV) lift
setpoint tolerance from 1 percent to 3
percent. In addition, the proposed change revises SR 3.1.7.10 to
increase the enrichment of sodium pentaborate used in the Standby
Liquid Control (SLC) system from >= 30.0 atom percent boron-10 to >=
45.0 atom percent boron-10.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change increases the allowable as-found MSSV lift
setpoint tolerance, determined by test after the valves have been
removed from service, from 1% to 3%. The
proposed change does not alter the TS requirements for the number of
MSSVs required to be operable, the nominal lift setpoints, the
allowable as-left lift setpoint tolerance, the MSSV testing
frequency, or the manner in which the valves are operated.
Consistent with current TS requirements, the proposed change
continues to require that the MSSVs be adjusted to within 1% of their nominal lift setpoints following testing. Since
the proposed change does not alter the manner in which the valves
are operated, there is no significant impact on reactor operation.
The proposed change does not involve a physical change to the
valves, nor does it change the safety function of the valves. The
proposed TS revision involves no significant changes to the
operation of any systems or components in normal or accident
operating
[[Page 4308]]
conditions and no changes to existing structures, systems, or
components, with the exception of the SLC system enrichment change.
The proposed change to increase the enrichment of sodium pentaborate
used in the SLC system will ensure that the requirements of 10 CFR
50.62, ``Requirements for reduction of risk from anticipated
transients without scram (ATWS) events for light-water-cooled
nuclear power plants,'' continue to be met. The SLC system is not an
initiator to an accident; rather, the SLC system is used to mitigate
an ATWS event. Therefore, these changes will not increase the
probability of an accident previously evaluated.
Generic considerations related to the change in setpoint
tolerance were addressed in NEDC-3175310, ``BWROG In-Service
Pressure Relief Technical Specification Revision Licensing Topical
Report,'' and were reviewed and approved by the NRC in a safety
evaluation dated March 8, 1993. General Electric Company (GE)
completed plant-specific analyses to assess the impact of the
setpoint tolerance increase on Dresden Nuclear Power Station Units 2
and 3 and QCNPS [Quad Cities Nuclear Power Station] Units 1 and 2.
The impact of the MSSV setpoint tolerance increase, as addressed in
this analysis, included vessel overpressure, Updated Final Safety
Analysis Report (UFSAR) Chapter 15 events, ATWS, Loss of Coolant
Accident (LOCA), containment response and loads, high pressure
systems performance, Appendix R fire protection, vessel thermal
cycle, operating mode and equipment out of service review, and
extended power uprate evaluation review. The proposed change to 3%
setpoint tolerance is supported by Westinghouse SVEA-96 Optimal fuel
analysis of events that credit the MSSVs.
The plant specific evaluations, required by the NRC's safety
evaluation and performed to support this proposed change, show that
there is no change to the design core thermal limits and adequate
margin to the reactor vessel pressure limits using a 3%
lift setpoint tolerance. These analyses also show that operation of
Emergency Core Cooling Systems is not affected, and the containment
response following a LOCA is acceptable. The plant systems
associated with these proposed changes are capable of meeting
applicable design basis requirements and retain the capability to
mitigate the consequences of accidents described in the UFSAR. The
accident analyses that credit the initiation of SLC as a dose
mitigation feature are not impacted by the proposed change because
the chemical properties of the SLC boron solution are not affected.
Therefore, these changes do not involve an increase in the
consequences of an accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change increases the allowable as-found lift
setpoint tolerance for the QCNPS MSSVs, and increases the required
enrichment of sodium pentaborate used in the SLC system. The
proposed change to increase the enrichment of sodium pentaborate
used in the SLC system will ensure that the requirements of 10 CFR
50.62 continue to be met.
The proposed change to increase the MSSV tolerance was developed
in accordance with the provisions contained in the NRC safety
evaluation for NEDC-31753P. MSSVs installed in the plant following
testing or refurbishment will continue to meet the current tolerance
acceptance criteria of 1% of the nominal setpoint. The
proposed change does not affect the manner in which the overpressure
protection system is operated; therefore, there are no new failure
mechanisms for the overpressure protection system. The proposed
change to allow an increase in the MSSV setpoint tolerance does not
alter the nominal MSSV lift setpoints or the number of MSSVs
currently required to be operable by QCNPS TS. The proposed change
does not involve physical changes to the valves, nor does it change
the safety function of the valves. There is no alteration to the
parameters within which the plant is normally operated. As a result,
no new failure modes are being introduced.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established through the design of the
plant structures, systems, and components, the parameters within
which the plant is operated, and the establishment of the setpoints
for the actuation of equipment relied upon to respond to an event.
The proposed change does not modify the safety limits or setpoints
at which protective actions are initiated, and does not change the
requirements governing operation or availability of safety equipment
assumed to operate to preserve the margin of safety.
Establishment of the 3% MSSV setpoint tolerance
limit does not adversely impact the operation of any safety-related
component or equipment. Evaluations performed in accordance with the
NRC safety evaluation for NEDC-31753P have concluded that all design
limits will continue to be met.
The proposed change to increase the enrichment of sodium
pentaborate used in the SLC system will ensure that the requirements
of 10 CFR 50.62 continue to be met.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
Based upon the above, EGC [Exelon Generation Company] concludes
that the proposed amendment presents no significant hazards
consideration under the standards set forth in 10 CFR 50.92 (c),
and, accordingly, a finding of no significant hazards consideration
is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Assistant General
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett
Square, PA 19348.
NRC Branch Chief: Michael L. Marshall, Jr.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: December 20, 2006.
Description of amendment request: The proposed amendment would
remove annotations referencing Technical Data Book (TDB)-VIII,
``Equipment Operability Guidance,'' and annotations referencing
Technical Specification Interpretations (TSIs) from the NRC Authority
File. These documents are used by Omaha Public Power District (OPPD)
personnel for additional guidance in applying certain Limiting
Conditions for Operation requirements to specific equipment and/or
situations. OPPD has annotated references to these documents in the
Technical Specification (TS) copies used at Fort Calhoun Station (FCS);
however, the annotations are ``pointers'' to additional guidance and
are not officially a part of the FCS TS. The proposed amendment also
corrects an administrative discrepancy in TS 2.10.4(1)(c).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The correction of administrative discrepancies in the Fort
Calhoun Station (FCS) Technical Specifications (TS) is not an
initiator of any previously evaluated accident. The proposed changes
will not prevent safety systems from performing their accident
mitigation function as assumed in the safety analysis.
Therefore, this change does not involve an increase in the
probability or consequences of any accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes only affect the Technical Specifications
and do not involve a physical change to the plant. Modifications
[[Page 4309]]
will not be made to existing components nor will any new or
different types of equipment be installed. This change will not
alter assumptions made in safety analysis and licensing bases.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The correction of administrative discrepancies in the Technical
Specifications has no impact on any safety analysis assumptions and
thus this TS change does not involve a reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn,
1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: David Terao.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: December 20, 2006.
Description of amendment request: The proposed amendment would
delete the Technical Specification (TS) requirements related to the
hydrogen purge system in TS 2.6(3) and TS Table 3-5, Item 17. The
proposed TS changes support implementation of the revisions to 10 CFR
50.44, ``Standards for Combustible Gas Control System in Light-Water-
Cooled Power Reactors,'' that became effective on September 16, 2003.
The changes are consistent with Revision 1 of NRC-approved Industry/
Technical Specification Task Force (TSTF) Standard Technical
Specification Change Traveler, TSTF-447, ``Elimination of Hydrogen
Recombiners and Change to Hydrogen and Oxygen Monitors.''
The NRC staff issued a notice of opportunity to comment in the
Federal Register dated August 2, 2002 (67 FR 50374), on possible
amendments for the elimination of requirements for hydrogen
recombiners, and hydrogen and oxygen monitors from the TSs, including a
model safety evaluation and model no significant hazards consideration
(NSHC) determination, using the consolidated line item improvement
process. The NRC staff subsequently issued a notice of availability of
the model for referencing in license amendment applications in the
Federal Register on September 25, 2003 (68 FR 55416). The licensee
affirmed the applicability of the NSHC in its application dated
December 20, 2006.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated.
The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen [and oxygen] monitors are no longer required to mitigate
design-basis accidents and, therefore, the hydrogen monitors do not
meet the definition of a safety-related component as defined in 10
CFR 50.2. RG 1.97 Category 1, is intended for key variables that
most directly indicate the accomplishment of a safety function for
design-basis accident events. The hydrogen [and oxygen] monitors no
longer meet the definition of Category 1 in RG 1.97. As part of the
rulemaking to revise 10 CFR 50.44 the Commission found that Category
3, as defined in RG 1.97, is an appropriate categorization for the
hydrogen monitors because the monitors are required to diagnose the
course of beyond design-basis accidents. [Also, as part of the
rulemaking to revise 10 CFR 50.44, the Commission found that
Category 2, as defined in RG 1.97, is an appropriate categorization
for the oxygen monitors, because the monitors are required to verify
the status of the inert containment.]
The regulatory requirements for the hydrogen [and oxygen]
monitors can be relaxed without degrading the plant emergency
response. The emergency response, in this sense, refers to the
methodologies used in ascertaining the condition of the reactor
core, mitigating the consequences of an accident, assessing and
projecting offsite releases of radioactivity, and establishing
protective action recommendations to be communicated to offsite
authorities. Classification of the hydrogen monitors as Category 3,
[classification of the oxygen monitors as Category 2] and removal of
the hydrogen [and oxygen] monitors from TS will not prevent an
accident management strategy through the use of the SAMGs, the
emergency plan (EP), the emergency operating procedures (EOP), and
site survey monitoring that support modification of emergency plan
protective action recommendations (PARs).
Therefore, the elimination of the hydrogen recombiner
requirements and relaxation of the hydrogen [and oxygen] monitor
requirements, including removal of these requirements from TS, does
not involve a significant increase in the probability or the
consequences of any accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility
of a New or Different Kind of Accident from any Previously
Evaluated.
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen [and oxygen] monitor requirements,
including removal of these requirements from TS, will not result in
any failure mode not previously analyzed. The hydrogen recombiner
and hydrogen [and oxygen] monitor equipment was intended to mitigate
a design-basis hydrogen release. The hydrogen recombiner and
hydrogen [and oxygen] monitor equipment are not considered accident
precursors, nor does their existence or elimination have any adverse
impact on the pre-accident state of the reactor core or post
accident confinement of radionuclides within the containment
building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety.
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen [and oxygen] monitor requirements,
including removal of these requirements from TS, in light of
existing plant equipment, instrumentation, procedures, and programs
that provide effective mitigation of and recovery from reactor
accidents, results in a neutral impact to the margin of safety.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The Commission has found that this
hydrogen release is not risk-significant because the design-basis
LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage.
Category 3 hydrogen monitors are adequate to provide rapid
assessment of current reactor core conditions and the direction of
degradation while effectively responding to the event in order to
mitigate the consequences of the accident. The intent of the
requirements established as a result of the TMI, Unit 2 accident can
be adequately met without reliance on safety-related hydrogen
monitors.
Therefore, this change does not involve a significant reduction
in the margin of safety. [The intent of the requirements established
as a result of the TMI, Unit 2 accident can be
[[Page 4310]]
adequately met without reliance on safety-related oxygen monitors.]
Removal of hydrogen [and oxygen] monitoring from TS will not result
in a significant reduction in their functionality, reliability, and
availability.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn,
1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: David Terao.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action, see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of application for amendments: December 27, 2004, as
supplemented by letters dated October 27, 2005, March 10, and October
6, 2006.
Brief description of amendments: The amendments revised the San
Onofre Nuclear Generating Station, Units 2 and 3, accident source term
used in the design-basis radiological consequence analyses. The
amendments were in accordance with the requirements of 10 CFR 50.67,
which addresses the use of an alternative source term (AST) at
operating reactors, and relevant guidance of Regulatory Guide (RG)
1.183. The amendments represent full-scope implementation of the AST
described in RG 1.183.
Date of issuance: December 29, 2006.
Effective date: As of the date of issuance and shall be implemented
within 180 days of issuance.
Amendment Nos.: Unit 2--210; Unit 3--202.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: February 1, 2005 (70 FR
5248). The supplemental letters dated October 27, 2005, March 10, and
October 6, 2006, provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the staff's original proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 29, 2006.
No significant hazards consideration comments received: No.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of application for amendments: March 10, 2006, as supplemented
by submittal dated May 16, 2006.
Brief description of amendments: The amendments conform the
Facility Operating Licenses NPF-10 and NPF-15 for the San Onofre
Nuclear Generating Station, Units 2 and 3 (SONGS 2 and 3) to reflect
their transfer from the City of Anaheim (Anaheim) to Southern
California Edison (SCE). The license transfers, which were approved by
the Order dated September 27, 2006, permitted the transfer of the 3.16-
percent undivided ownership interest in the facilities held by Anaheim
to SCE, excluding Anaheim's interest in its spent fuel and in the SONGS
2 and 3 independent spent fuel storage installation. SCE retains
exclusive responsibility and control over the operation of SONGS 2 and
3.
Date of issuance: December 29, 2006.
Effective date: At the time the transfer is completed.
Amendment Nos.: Unit 2--209; Unit 3--201.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Facility Operating Licenses.
Date of initial notice in Federal Register: June 8, 2006 (71 FR
33321) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 27, 2006.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-259 Browns Ferry Nuclear
Plant, Unit 1, Limestone County, Alabama
Date of application for amendment: January 6, 2006 (TS-443), as
supplemented by letter dated October 2, 2006.
Brief description of amendment: Activation of thermal-hydraulic
stability monitoring instrumentation. The Oscillation Power Range
Monitor System is designed to provide the licensee's solution regarding
reactor stability.
Date of issuance: December 29, 2006.
Effective date: Date of issuance, to be implemented within 60 days.
Amendment No.: 266.
Renewed Facility Operating License No. DPR-33: Amendment revised
the TSs.
Date of initial notice in Federal Register: April 5, 2006 (71 FR
23962). The October 2, 2006, supplement, contained clarifying
information and
[[Page 4311]]
did not change the NRC staff's initial proposed finding of no
significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated December 29, 2006.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room (PDR), located at One White Flint North, Public File Area 01F21,
11555 Rockville Pike (first floor), Rockville, Maryland. Publicly
available records will be accessible from the Agencywide Documents
Access and Management System's (ADAMS) Public Electronic Reading Room
on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems
in accessing the documents located in ADAMS, contact the PDR Reference
staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to [email protected].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. Within 60 days after the date
of publication of this notice, the licensee may file a request for a
hearing with respect to issuance of the amendment to the subject
facility operating license and any person whose interest may be
affected by this proceeding and who wishes to participate as a party in
the proceeding must file a written request for a hearing and a petition
for leave to intervene. Requests for a hearing and a petition for leave
to intervene shall be filed in accordance with the Commission's ``Rules
of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2.
Interested persons should consult a current copy of 10 CFR 2.309, which
is available at the Commission's PDR, located at One White Flint North,
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland, and electronically on the Internet at the NRC Web site,
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR Reference staff at
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. If a
request for a hearing or petition for leave to intervene is filed by
the above date, the Commission or a presiding officer designated by the
Commission or by the Chief Administrative Judge of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the Chief Administrative Judge of the Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the petitioner/
[[Page 4312]]
requestor seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
intends to rely to establish those facts or expert opinion. The
petition must include sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or
fact.\1\ Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A petitioner/requestor
who fails to satisfy these requirements with respect to at least one
contention will not be permitted to participate as a party.
---------------------------------------------------------------------------
\1\ To the extent that the applications contain attachments and
supporting documents that are not publicly available because they
are asserted to contain safeguards or proprietary information,
petitioners desiring access to this information should contact the
applicant or applicant's counsel and discuss the need for a
protective order.
---------------------------------------------------------------------------
Each contention shall be given a separate numeric or alpha
designation within one of the following groups:
1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the
applications.
2. Environmental--primarily concerns/issues relating to matters
discussed or referenced in the environmental analysis for the
applications.
3. Miscellaneous--does not fall into one of the categories outlined
above.
As specified in 10 CFR 2.309, if two or more petitioners/requestors
seek to co-sponsor a contention, the petitioners/requestors shall
jointly designate a representative who shall have the authority to act
for the petitioners/requestors with respect to that contention. If a
petitioner/requestor seeks to adopt the contention of another
sponsoring petitioner/requestor, the petitioner/requestor who seeks to
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the
petitioners/requestors with respect to that contention.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First-class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer or the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
STP Nuclear Operating Company, Docket No. 50-498, South Texas Project,
Unit 1, Matagorda County, Texas
Date of amendment request: December 20, 2006, as supplemented by
letter dated December 28, 2006.
Description of amendment request: The amendment, for a one-time
change, revised Technical Specification (TS) 3.3.2 for the loss of
power (LOP) instrumentation (Functional Unit 8, ``loss of power'') in
TS Table 3.3-3, ``Engineered Safety Features Actuation System
Instrumentation.'' A note is added to TS Table 3.3-3, Action 20, which
is the TS-required action for inoperable LOP instrumentation, to allow
a one-time provision for corrective maintenance on an inoperable Unit 1
LOP instrumentation channel when the number of operable channels are
more than one less than the total number of channels. This provision
for corrective maintenance expires 30 days after the amendment is
approved.
Date of issuance: January 11, 2007.
Effective date: Effective as of its date of issuance and shall be
implemented by January 15, 2007.
Amendment No.: 176.
Facility Operating License No. NPF-76: The amendment revised the
Technical Specifications and Facility Operating License.
Public comments requested as to proposed no significant hazards
consideration (NSHC): No.
The Commission's related evaluation of the amendment, finding of
emergency circumstances, state consultation, and final NSHC
determination are contained in a safety evaluation dated January 11,
2007.
Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Branch Chief: David Terao.
Dated at Rockville, Maryland, this 22nd day of January 2007.
For the Nuclear Regulatory Commission.
John W. Lubinski,
Deputy Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. E7-1259 Filed 1-29-07; 8:45 am]
BILLING CODE 7590-01-P